This invention relates to a novel zirconium based alloy plate, a method of manufacturing this alloy plate, a fuel channel and a fuel assembly using this alloy plate.
Zirconium is a material having high corrosion resistance and a small neutron absorption cross section and is therefore used for a reactor fuel assembly member. For this kind of use, Zr-Sn-Fe-Cr-Ni alloys called zircaloy-2 and zircaloy-4 are mainly used. If these alloys are used in a nuclear reactor for a long time, elongation, bending and deformation in particular directions occur as shown in FIG. 2, because (0001) planes thereof are oriented in the direction of plate thickness. If a bending deformation occurs in a fuel channel box, the space for driving the control rod is reduced. Such a phenomenon impedes the operation of the reactor. Also, if a bending deformation occurs, the distance to the fuel cladding tube is changed so that the ratio of water and uranium is locally changed, thereby causing a change in fission reactivity and, hence, abnormal heating. By this effect, corrosion of the fuel cladding tube is accelerated. Further, fuel damage may thereby be caused. To prevent bending deformation of the fuel channel box due to such non-uniformity of neutron exposure, uniformization of neutron exposure based on changing the fuel assembly loading position in the reactor core has been studied. This study, however, has not succeeded in preventing bending deformation. The reduction in the control rod driving space and the change in fission reactivity caused by bending deformation are major factors of limitation of the service life of the fuel channel box.
The corrosion of the fuel channel box also is a factor of limitation of the service life. Japanese Patent Examined Publication Nos. 56-12310 and 60-44387 disclose a method for improving the corrosion resistance by a heat treatment based on quenching a Zr alloy from an .alpha.+.beta. phase temperature range or a .beta. phase temperature range. By this method, however, the bending deformation due to irradiation growth cannot be reduced for the reason described later, and this method provides no effective technique for limiting irradiation growth of zirconium alloy members.
This conventional method uses a heat treatment for improving only the corrosion resistance of zirconium alloy members without changing the grain size, orientation and mechanical characteristics thereof. For this heat treatment, therefore, heating to an .alpha.+.beta. phase temperature range (800 to 980.degree. C.) rather than a .beta. phase temperature range (.gtoreq.980.degree. C.) in which coarsening of crystal grains occurs, followed by quenching, is effected. The crystal orientation, which is an important factor of limitation of irradiation growth, is not changed by this heat treatment and therefore no effective irradiation growth limiting technique is provided. Japanese Patent Unexamined Publication No.59-229475 discloses a method of orienting crystals so that an fl value is 0.15 to 0.5. This method also provides no irradiation growth limiting technique or rather promotes irradiation growth.